ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller & Associates I MJW Corporation

Page 1 of 29

Document Title:

Paducah Gaseous Diffusion Plant – Site 

Description

Document Number: ORAUT-TKBS-0019-2 

Revision: 01

Effective Date: 08/21/2006

Type of Document: TBD

Supersedes: Revision 00

Subject Expert: Baynard Turpin

Document Owner

Approval: Signature on File Approval Date: 07/11/2006

Daniel Mantooth, TBD Team Leader

Approval: Signature on File Approval Date: 07/12/2006

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 07/11/2006

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 07/27/2006

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 08/21/2006

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0019-2 Revision No. 01 Effective Date: 08/21/2006 Page 2 of 29

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

05/07/2004 00 New technical basis document for the Paducah Gaseous Diffusion 

plant – Site Description. First approved issue. Initiated by Jay J. 

Maisler.

08/21/2006 01 This revision addresses comment #2 from the meeting with United 

Steelworkers Local 5-550 (formerly PACE) and SPFPA Local 111 

held on 02/10/2005. Approved issue of Revision 01. This revision 

results in no change to the assigned dose and no PER is required. 

Training required: As determined by the Task Manager. Initiated by 

Paul A. Szalinski.
Document No. ORAUT-TKBS-0019-2 Revision No. 01 Effective Date: 08/21/2006 Page 3 of 29

TABLE OF CONTENTS

SECTION TITLE

Acronyms and Abbreviations ............................................................................................................... 5

PAGE

2.1 Introduction .............................................................................................................................. 6

2.1.1 Purpose......................................................................................................................... 7

2.1.2 Scope............................................................................................................................ 7

2.2 Site Description........................................................................................................................ 7

2.3 Site Activities............................................................................................................................ 7

2.4 Site Processes .......................................................................................................................12

2.4.1 Key Uranium Processing Facilities ..............................................................................12

2.4.2 C-310 – Purge and Product Withdrawal Building.........................................................15

2.4.3 C-315 – Surge and Tails Withdrawal Building..............................................................15

2.4.4 Gaseous Diffusion Process Buildings (C-331, C-333, C-335, and C-337) ................... 16

2.4.5 Building C-340 – Metals Plant .....................................................................................17

2.4.6 Building C-400 – Decontamination and Cleaning Building/Uranium Recovery............. 19

2.4.7 Building C-404 – Solid Radioactive Waste Disposal Area............................................ 21

2.4.8 Building C-409 – Stabilization Building ........................................................................21

2.4.9 Building C-410 – Feed Plant........................................................................................22

2.4.10 Building C-420 – Oxide Conversion Plant (UF4 – Green Salt Plant)............................. 23

2.4.11 Building C-710 – Analytical Laboratories .....................................................................23

2.4.12 Building C-720 – Maintenance Building .......................................................................24

References ........................................................................................................................................26

Glossary ............................................................................................................................................27
Document No. ORAUT-TKBS-0019-2 Revision No. 01 Effective Date: 08/21/2006 Page 4 of 29

LIST OF TABLES

TABLE TITLE

2-1 Buildings and potential exposures to recycled uranium ..........................................................

PAGE

10

2-2 Average external cumulative penetrating dose (deep dose equivalent) and 

number of workers assigned to each department, 1953–1988 ............................................... 11

2-3 Average recorded external radiation doses (deep dose equivalent) per worker 

per year from 1953 to 1988, including maximum exposures recorded for any 

single worker for that year...................................................................................................... 11

2-4 Processes and facilities involving potential worker exposure.................................................. 13

LIST OF FIGURES

FIGURE TITLE

2-1 PGDP area map.......................................................................................................................

PAGE

7

2-2 PGDP site map ........................................................................................................................8

2-3 PGDP uranium enrichment process .........................................................................................9
Document No. ORAUT-TKBS-0019-2 Revision No. 01 Effective Date: 08/21/2006 Page 5 of 29

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

CIP Cascade Improvement Program

CUP Cascade Upgrade Program 

d day

DOE U.S. Department of Energy

EEOICPA Energy Employees Occupational Illness Compensation Program Act

ft foot

g gram

HF hydrofluoric acid

kg kilogram

KOW Kentucky Ordnance Works

L liter

lb pound

MgF2 magnesium fluoride

mg milligram

mrem millirem

ORNL Oak Ridge National Laboratory

PEMU Process Equipment Modification and Uprating Program

PGDP Paducah Gaseous Diffusion Plant

POC probability of causation

ppb parts per billion

ppm parts per million

RU recycled uranium

TBD technical basis document

TRU transuranic

UF4 uranium tetrafluoride (“green salt”)

UF6 uranium hexafluoride 

UO2 uranium dioxide

UO2F2 uranyl fluoride

UO3 uranium trioxide (yellowcake) 

U3O8 triuranium octaoxide 

U.S.C. United States Code
Document No. ORAUT-TKBS-0019-2 Revision No. 01 Effective Date: 08/21/2006 Page 6 of 29

2.1 INTRODUCTION

Technical basis documents (TBDs) and site profile documents are general working documents that 

provide guidance concerning the preparation of dose reconstructions at particular sites or categories 

of sites. They will be revised in the event additional relevant information is obtained about the 

affected site(s). These documents may be used to assist the National Institute for Occupational 

Safety and Health (NIOSH) in the completion of the individual work required for each dose 

reconstruction.

In this document, the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy [DOE] facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 7384l(5) and (12)]. 

EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds upon 

which such building, structure, or premise is located ... in which operations are, or have been, 

conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

above, any facility that performs or performed DOE operations of any nature whatsoever is a DOE 

facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at 

least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the POC [probability of causation1

As noted above, the statute includes a definition of a DOE facility that excludes “buildings, structures, 

premises, grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982

(42 U.S.C. 7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 7384l(12)]. 

While this definition contains an exclusion with respect to the Naval Nuclear Propulsion Program, the 

section of EEOICPA that deals with the compensation decision for covered employees with cancer 

[i.e., 42 U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all occupationally derived radiation 

exposures at the facility in its dose reconstructions for employees at DOE facilities, including radiation 

exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and external 

dosimetry monitoring results are considered valid for use in dose reconstruction. No efforts are made 

to determine the eligibility of any fraction of total measured exposure for inclusion in dose 

reconstruction. NIOSH, however, does not consider the following exposures to be occupationally 

derived:

] guidelines established under 

subsection (c) ...” [42 U.S.C. § 7384n(b)]. Neither the statute nor the POC guidelines (nor the dose 

reconstruction regulation) define “performance of duty” for DOE employees with a covered cancer or 

restrict the “duty” to nuclear weapons work.

• Radiation from naturally occurring radon present in conventional structures

• Radiation from diagnostic X-rays received in the treatment of work-related injuries

 

1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC. 
Document No. ORAUT-TKBS-0019-2 Revision No. 01 Effective Date: 08/21/2006 Page 7 of 29

Figure 2-1. PGDP area map.

2.1.1

The purpose of the Paducah Gaseous Diffusion Plant (PGDP) has been and continues to be the 

enrichment of uranium, initially for military applications and subsequently for commercial nuclear 

reactor fuel. PGDP enriches feed material in the form of uranium hexafluoride (UF6) gas from 

approximately 0.711% 235U up to about 2.5% 235U. The enriched product from PGDP was sent to 

other DOE gaseous diffusion plants at Portsmouth, Ohio, and Oak Ridge, Tennessee, for further 

enrichment.

Purpose

2.1.2

This TBD, which is part of the Paducah Gaseous Diffusion Plant (PGDP) Site Profile, provides the site 

description (Section 2.2), routine site activities (Section 2.3), and site processes by specific building 

(Section 2.4) most relevant to worker dose reconstruction.

Scope

2.2 SITE DESCRIPTION

PGDP was built on a portion of 16,126 acres of 

farmland acquired by the U.S. War Department 

[the U.S. Department of Defense did not exist 

until 1947] during World War II. The War 

Department acquired this land for a munitions 

facility, the Kentucky Ordnance Works (KOW), 

which was operated by Atlas Powder Company 

until it closed in 1946 (Lockheed Martin 1997a). 

The KOW included a trinitrotoluene 

manufacturing area; an acid production area; 

coal, sulfur, toluene, and ordnance storage 

areas; a sewage treatment plant; a water 

treatment plant; and burning grounds. PGDP 

still uses the water treatment plant. In 1950, 

7,556 acres of the land east of the KOW were 

acquired by the U.S. Atomic Energy 

Commission (AEC; a DOE predecessor agency) 

as a site for a uranium enrichment facility that 

became PGDP (see Figure 2-1). The plant 

began operating in 1952, but construction was 

not complete until 1954. The facility reservation 

covered 3,424 acres, with about 750 acres inside the security fence. The rest of the land was 

transferred to the Tennessee Valley Authority for the Shawnee Steam Plant and to the 

Commonwealth of Kentucky for wildlife conservation and recreational purposes (Lockheed Martin 

1997b). The site consisted of 161 buildings, four of which contain the gaseous diffusion process (see 

Figure 2-2).

2.3 SITE ACTIVITIES

Originally, most UF6 feed material came from the depleted tails produced during normal diffusion 

operations at PGDP and from the Oak Ridge and Portsmouth gaseous diffusion plants. In addition, 

from 1953 through 1977, UF6 feed material was produced from uranium trioxide (UO3), called 

yellowcake, at PGDP in Buildings C-410 and C-420. The UO3 was supplied by sources such as El 

Dorado Mining and Refining, Mallinckrodt Chemical Works, and General Chemicals (now Allied 
Document No. ORAUT-TKBS-0019-2 Revision No. 01 Effective Date: 08/21/2006 Page 8 of 29

Chemical), and comprised less than 10% of the UF6 fed to the 

cascade. From 1953 through 1964 and intermittently from 

1968 through 1977, the feed plant produced UF6 from UO3

from spent reactor fuel processed at the DOE Hanford and 

Savannah River Sites. After 1977, all feed came in the form of 

UF6 from outside sources such as Oak Ridge, Portsmouth, 

and Allied Chemical (Bechtel Jacobs 2001).

Feed material was made from production reactor tails from 

1953 until 1964 and intermittently from 1968 to 1977. The 

percentage of PGDP cascade feed material from reactor tails 

averaged 17% during the periods this material was used, 

ranging from 3% in 1975 to 65% in 1973. Processing of UO3

to UF6 occurred in three steps: reduction, hydrofluorination, 

and fluorination. The uranium exposure pathway of greatest 

concern at PGDP was inhalation of uranium dust. 

Chemical reduction involved transforming UO3 into uranium 

dioxide (UO2) using hydrogen gas. Hydrofluorination of UO2

into uranium tetrafluoride (UF4), commonly referred to as 

green salt, was accomplished by adding anhydrous 

hydrofluoric acid (HF). Fluorination occurred in Building C-410 using heated elemental fluorine gas in 

tower reactors. The first two steps were performed on vibration tray reactors (shaker trays) from 1953 

to 1956. In 1956, due to frequent equipment failures, spills, leaks, and the increased demand for 

feed, Building C-420, commonly called the green salt plant, was completed and green salt production 

at Building C-410 was phased out. In Building C-420, reduction was performed in two-stage fluidized 

bed reductors; hydrofluorination was performed in three sets of horizontal screw reactors or in a twostage fluidized bed hydrofluorinator. High Radiation Areas existed near the fluorination towers and 

ash receivers.

The main process buildings at PGDP (C-331, C-333, C-335, and C-337) contain the cascades, which 

are a series of compressor and converter stages and supporting equipment arranged in units and 

cells that progressively enrich UF6 in its gaseous form. Enrichment occurs as UF6 passes through 

barriers in the converters allowing isotopes of lower molecular weight to pass through. The series of 

converters results in two streams of UF6 – one of progressively higher percentage 235U that moves to 

the product withdrawal station in Building C-310, and one of progressively higher percentage 238U that 

moves toward the tails withdrawal station in Building C-315. Both the enriched product and the 

depleted tails are fed as liquid into cylinders and allowed to cool until solid. The enriched product is 

shipped to Portsmouth for further enrichment. The depleted material either was re-fed to the cascade 

or is stored on the site. Figure 2-3 shows the uranium enrichment process.

There were two cascade improvement/upgrade programs at PGDP – the Cascade Improvement 

Program (CIP) and the Cascade Upgrade Program (CUP). The first ran from 1958 to 1962 and the 

second from 1973 to 1981. These programs were significant because of possible worker exposure to 

transuranic (TRU) elements while the cascade systems were open. Other major events were the 

closing of feed plant operations from 1965 to 1969 and in 1971. Feed plant operations, in 

Buildings C-410 and C-420 and the decontamination building, C-400, were permanently shut down in 

the late 1970s.

PGDP processed recycled uranium (RU) from initial startup in fiscal year 1953 through fiscal year 

1989 (excluding fiscal years 1965 through 1968, 1976, and 1977 through 1985). In 1957, the 

Figure 2-2. PGDP site map.
Document No. ORAUT-TKBS-0019-2 Revision No. 01 Effective Date: 08/21/2006 Page 9 of 29

Figure 2-3. PGDP uranium enrichment process (Lockheed Martin 1997a).

presence of 237Np and 99Tc was documented; between 1959 and 1966, PGDP and the AEC 

conducted studies related to the behavior, health effects, and controls for these radionuclides. The 

concentration of TRU elements, such as 237Np and 239Pu, and fission products, and 99Tc, in the reactor 

tails material was small, estimated at approximately 0.2 ppm neptunium, 4 ppb 239Pu, and 7 ppm 99Tc. 

In addition, 241Am, a decay product of 239Pu, builds up as 239Pu decays. These radionuclides were 

concentrated during processing at specific locations, which could increase radiation exposure to

certain workers in these areas (Bechtel Jacobs 2001). Table 2-1 lists assessments of activities that 

are assumed to have probably involved potential worker exposure to RU constituents. In some cases, 

concentrations for 99Tc were not provided in the source reference (Bechtel Jacobs 2001) and are

listed as “XX.”

Based on exposure records in the database, worker interviews, and health physics and inspection 

reports, an estimated 2,500 to 4,000 workers worked in areas with “moderate” to “high” potential for 

increased internal and external radiation exposures. This is based on a relative ranking of the 

potential of radiation exposures at PGDP. These areas included the feed plant (Buildings C-410 and 

C-420) with operators and mechanics receiving the highest doses, respectively; the decontamination 

building (C-400) with decontamination workers receiving the highest doses; and the cascade buildings 

(C-331, C-333, C-335, and C-337) with operators receiving the highest doses. In addition, workers in 
Document No. ORAUT-TKBS-0019-2 Revision No. 01 Effective Date: 08/21/2006 Page 10 of 29

Table 2-1. Buildings and potential exposures to recycled uranium.

a

Location Activity Time Period

Maximum 

radionuclide 

concentrations 

expected

Radiological 

risk

C-333

C-337

C-410

Cascade operations–Operators could have been 

exposed to gaseous UF6 when connecting or 

disconnecting feed cylinders containing recycled uranium.

1953–1964

1969–1970

1972–1976

450 ppb Np

0.09 ppb Pu

23,000 Tc-99

(UF6)

2 

Moderate

C-331

C-333

C-335

C-337

C-400

C-409

C-720

Cascade maintenance–Maintenance workers could have 

been exposed to constituents of interest when working on 

or removing interior surfaces of cascade equipment near 

recycled uranium feed points (cascade dust). 

1954–1961

1973–1981

(CIP/CUP)

2740 ppb Pu

3,220,000 ppb Np

XX Tc-99

(cascade dust)

3

High

C-410

C-420

Feed plant operations, RU to UF6–

UO3 → UO2 UF4 UF 

(handling, drumming, bag changing, etc.) (tower ash)

1953–1964

1969–1977

1982–1983

(shipped)

25,602 ppb Np

8,000 ppb Pu

4,600 ppb Tc-99

(tower ash)

3

High

C-400

C-710

Uranium/neptunium recovery–Workers salvage uranium 

from cleaning solutions and waste products (cylinder 

wash, ash, dust). Aqueous solutions containing >2 mg/L 

Np were reprocessed, <2 mg/L were discarded.

1958 to late 

1970’s

25,000,000 ppb Np

100,000 ppb Pu

XX ppb Tc-99 

(solutions)

2 

Moderate

C-340 Production and handling of UF4 produced from tails–

UF6, UF6 → UF4

1957–1962

1967–1977

62.6 ppb Np

0.063 ppb Pu

49 ppb Tc-99

(UF4)

1

Low

C-405 Handling ash from contaminated items–Incinerator 

(filter bags, etc.)

1953–1990 3,512 ppb Np

56 ppb Pu

921 ppb Tc-99

(ash)

3

High

C-315

C-340

Connecting and disconnecting UF6 tails cylinders and 

handling UF6 tails

1953–1999 5 ppb Np

0.1 ppb Pu

20,000 ppb Tc-99

(UF6)

1

Low

C-310 Connecting and disconnecting UF6 product cylinders 

and handling UF6 product

1953–1999 27 ppb Np

0.06 ppb Pu

20,000 ppb Tc-99

(UF6)

1

Low

C-410

C-310

Changing/cleaning MgF2 traps 1964–1966 1,900 ppb Np

0.01 ppb Pu

2,266,000 ppb Tc-99

(MgF2)

2 

Moderate

C-340 Uranium metal production from metal production, 

sawing, and metal handling

1957–1962

1967–1977

2.7 ppb Np

0.0085 ppb Pu

10.6 ppb Tc-99

(U metal)

1

Low

C-340 Handling MgF2 in uranium metal manufacturing–

Cleaning, roasting, liner preparation, knock out, and 

drumming.

1957–1962

1968–1977

2,400 ppb Np

1.2 ppb Pu

1,880 ppb Tc-99

(MgF2) 

3

High

Source: Bechtel Jacobs (2001, Table 2.4-1) 

Buildings C-340 and C-720 had a potential for increased external radiation exposures. Table 2-2 lists 

average doses per department (for 1953–1988) (PACE/University of Utah 2000).

Data indicate that approximately 200 individuals received more than 1 rem external exposure in any 

one calendar year. Health physics reports documented that many workers exceeded the weekly plant 

action level of 300 mrem/week, but suggested that PGDP kept workers below the annual regulatory 

limits by rotation of duties and limitations on stay time in areas of higher exposure. Table 2-2 lists the 

average external cumulative penetrating dose (deep dose equivalent) and number of workers 
Document No. ORAUT-TKBS-0019-2 Revision No. 01 Effective Date: 08/21/2006 Page 11 of 29

Table 2-2. Average external cumulative penetrating dose (deep dose equivalent) and number of 

workers assigned to each department, 1953–1988. 

Dept. 

No.

Job 

description

Average 

cumulative 

dose 

(mrem)

Number 

of 

workers

5751 Feed plant operators 3,814 185

5760 Decontamination 2,788 116

5034 Feed plant mechanics 2,587 99

5015 Unknown 2,025 17

5676 Unknown 861 14

5730 Cascade operators 627 578

5785 Chemical operators 595 113

5075 Instrumentation 538 245

5020 Unknown 481 17

5008 Transportation pool 371 33

5002 Process maintenance 364 578

5108 Environmental control 338 48

5268 Unknown 316 236

5077 Electricians 298 318

5005 Material termination mgr. 295 90

5772 PEMU decontamination 253 22

5759 Unknown 220 4

Dept. 

No.

Job 

description

Average 

cumulative 

dose 

(mrem)

Number 

of 

workers

5049 Unknown 182 12

5725 Unknown 175 20

5044 Mechanical inspection 170 113

5021 Plant services 147 486

5770 Converter test 145 23

5035 Feed plant mechanics 143 160

5019 Unknown 142 13

5740 Nitrogen plant 142 22

5646 Metals building 132 95

5674 Unknown 129 8

5048 Fabrication shops 127 667

5023 Unknown 115 24

5675 Unknown 114 7

5743 Steam plant 111 61

5027 Unknown 110 282

Total 4,706

assigned to each department, 1953–1988 (PACE/University of Utah 2000), while Table 2-3 lists 

average and maximum doses (PACE/University of Utah 2000).

Table 2-3. Average recorded external radiation doses (deep dose equivalent) per worker per year 

from 1953 to 1988, including maximum exposures recorded for any single worker for that year. 

Year

Average 

recorded dose 

(mrem)a

Maximum 

recorded 

dose (mrem)

Number 

of 

workers

1953 139.8 820 223

1954 232.5 1,580 284

1955 241.9 2,500 417

1956 358.6 4,700 471

1957 251.7 3,190 669

1958 185.3 3,630 661

1959 201.5 2,360 570

1960 201.1 2,510 526

1961 177.0 2,530 1,690

1962 149.5 2,980 1,479

1963 144.1 3,040 1,311

1964 73.4 1,860 1,219

1965 34.1 1,610 1,128

1966 31.7 1,470 1,138

1967 49.8 1,120 1,143

1968 61.8 1,400 1,241

1969 73.3 1,970 1,270

1970 41.7 840 1,273

Year

Average 

recorded dose 

(mrem)a

Maximum 

recorded 

dose (mrem)

Number 

of 

workers

1971 62.4 1,380 1,254

1972 58.9 1,760 1,288

1973 53.0 1,830 1,404

1974 26.5 1,030 1,624

1975 50.1 1,049 2,013

1976 35.1 1,224 2,426

1977 23.2 742 2,643

1978 39.9 359 2,613

1979 8.2 364 2,487

1980 18.2 344 2,308

1981 7.6 420 1,840

1982 6.5 350 1,617

1983 6.7 340 1,452

1984 9.2 420 1,434

1985 6.1 350 1,365

1986 9.6 490 1,244

1987 8.0 470 1,275

1988 6.5 720 1,359

a. The large number of zero values in the database reduces average values for radiation workers.

Approximately 10% of the 2,500 to 4,000 workers had a potential for higher-than-average radiation 

exposures. Reports that indicate extensive radioactive contamination in lunchroom areas, workers 

covered with black soot after ash-handling operations, Building C-410 floors routinely covered with 

visible green powder, and surveys indicating elevated concentrations of transuranics in almost all 
Document No. ORAUT-TKBS-0019-2 Revision No. 01 Effective Date: 08/21/2006 Page 12 of 29

process buildings illustrate inadequacies in the PGDP contamination control program (PACE/

University of Utah 2000).

In general, job categories that had increased potential for exposures from external radiation were 

similar to those for internal radiation. The jobs and tasks that appear to involve the greatest potential 

for radiation exposure included ash handling (Building C-410), cylinder heel cleaning (Building C-400), 

derby processing (Building C-340), pulverizer operations (Building C-400), certain maintenance 

operations on the fluorination towers (Building C-410), maintenance on cascade equipment 

(cascades), cleaning of air filters (baghouses for Buildings C-400, C-410, C-420, and C-340), 

converter maintenance (Building C-720), flange grinding (Buildings C-340, C-400, C-410, and C-420), 

maintenance of the hydrogenation towers (Building C-340), and decontamination building cleaning 

operations (Building C-400). Jobs and tasks classified as having moderate potential radiation 

exposures include cascade operators and instrument mechanics (cascades), green salt sweeping 

(Buildings C-410 and C-420), disassembly of compressor (Building C-720), disassembly of block 

valves (Building C-720), drumming of green salt (Building C-340), and baghouse cleaning for 

cascades (Buildings C-310 and C-315). Other jobs and tasks had lower potential for worker radiation 

exposure. In addition, due to the lack of aggressive contamination control programs, contamination 

from the operations discussed above presented potential exposures to surrounding work areas. 

Further, legacy contamination generated from these operations could have posed a potential radiation 

exposure to individuals with access to the PGDP site in later years.

Special Incidences or Activities – High Potential for Increased Radiation Exposure

C-310 fire, 1956

C-337 fire, 1962

C-340 explosion and fire, 1962

Two workers overexposed, 1968, unknown location

C-315 fire, 1978

Cascade Improvement Program, 1954–1962

Cascade Improvement Program, 1973–1981

C-400 neptunium production

2.4 SITE PROCESSES

2.4.1

The major facilities at PGDP are:

Key Uranium Processing Facilities

• C-300 – Central Control Building

• C-310 – Purge and Product Withdrawal Building

• C-315 – Surge and Tails Withdrawal Building

• C-331, C-333, C-335, and C-337 – Gaseous Diffusion Process Buildings

• C-340 – Metals Plant (currently inactive)

• C-400 – Decontamination and Cleaning Building

• C-410/C-420 – UF6 Feed Plant (currently inactive)

This section includes other key facilities where moderate personnel radiation exposure occurred (e.g., 

C-404, C-409, C-710, and C-720). Table 2-4 lists the processes and facilities involving potential 

worker exposures (Bechtel Jacobs 2001).
Document No. ORAUT-TKBS-0019-2 Revision No. 01 Effective Date: 08/21/2006 Page 13 of 29

Table 2-4. Processes and facilities involving potential worker exposure. 

Process Facilities Time Period Potential RU exposure Notes

Unpacking, feeding of UO3 to 

feed production, reduction of 

UO3 to UO2, hydrofluorination 

to UF4.

C-420 1953 – 1964

1969 – 1977

RU constituents of concern 

not concentrated at this 

point in process.

Flame tower fluorination of 

UF4 to UF6, including 

generation of ash and filter 

residue.

C-410,

C-420

1953 – 1964

1968 – 1977

Pu and Np and some Tc-99 

were concentrated in ash 

and residue.

Venting to atmosphere from 

flame tower fluorination. 

C-410,

C-420

1953 – 1964

1969 – 1977

Venting did not involve 

concentrated RU.

Filters were potential 

point for concentration.

Collection of ash for U 

recovery and cleaning of 

tower filters. 

C-410,

C-420

1953 – 1964

1969 – 1977

Pu and Np and some Tc-99 

were concentrated in ash 

and residue.

Potential for worker 

exposure to 

concentrated RU 

constituents.

Pulverization of tower ash 

and other recovery 

processes. 

C-410,

C-420

1953 – 1964

1968 – 1976

Pu and Np and some Tc-99 

were concentrated in ash 

and residue.

Potential for worker 

exposure to 

concentrated RU 

constituents.

Recycling of U in ash through 

fluorination. 

C-410,

C-420

1953 – 1964

1969 – 1977

Pu and Np and some Tc-99 

were concentrated in ash 

and residue.

Potential for worker 

exposure to 

concentrated RU 

constituents.

Packing of waste ash for 

storage prior to shipping. 

C-410,

C-420

1953 – 1964

1969 – 1977

Pu and Np and some Tc-99 

were concentrated in ash 

and residue.

Potential for worker 

exposure to 

concentrated RU 

constituents.

Transfer of UF6 to cylinder 

from fluorination process. 

C-410,

C-420

1953 – 1964

1969 – 1977

Pu and Np and some Tc-99 

became further 

concentrated in cylinder 

heels.

Withdrawing UF6 from 

cylinder and feeding into 

cascade. 

Cascade feed 

points

1953 – 1964

1969 – 1970

1972 - 1976

Pu and Np and some Tc-99 

were concentrated in 

cylinder heels. Np and 

smaller amounts of Pu 

exiting cylinder with UF6

collected at feeder head 

leading to cascade and in 

cascade near feed point.

Venting to atmosphere from 

cascade. 

Cascade 

purge 

locations; 

C-331, 

C-333, 

C-335, 

C-337, 

C-310

1952 - 1985 As RU moved higher in 

cascade, Tc-99 became 

concentrated, but Pu and 

Np were much diminished. 

Venting at higher points in 

cascade involved 

concentrated Tc-99.

Traps were used to 

remove Tc-99 before 

venting.

CIP/CUP equipment removal 

and other equipment removal, 

maintenance, and 

decontamination. 

Cascade 

buildings 

C-270 Scrap 

Yard Burial 

Grounds

1954 - 1962

1973 - 1981

Np and lesser amounts of 

Pu collected near cascade 

feed points.

Removal, handling, 

and decontamination 

of equipment near feed 

points could have 

exposed workers to 

concentrations of small 

percentages of Np and 

Pu that actually 

entered cascade.
Document No. ORAUT-TKBS-0019-2 Revision No. 01 Effective Date: 08/21/2006 Page 14 of 29

Table 2-4 (Continued). Processes and facilities involving potential worker exposure. 

Process Facilities Time Period Potential RU exposure Notes

Unpacking, feeding of UO3 to 

feed production, reduction of 

UO3 to UO2, hydrofluorination 

to UF4.

C-420 1953 – 1964

1969 – 1977 

RU constituents of concern 

not concentrated at this 

point in process.

Flame tower fluorination of 

UF4 to UF6, including 

generation of ash and filter 

residue.

C-410,

C-420

1953 – 1964

1968 – 1977

Pu and Np and some Tc-99 

were concentrated in ash 

and residue.

Venting to atmosphere from 

flame tower fluorination. 

C-410,

C-420

1953 – 1964

1969 – 1977

Venting did not involve 

concentrated RU.

Filters were potential 

point for concentration.

Collection of ash for U 

recovery and cleaning of 

tower filters. 

C-410,

C-420

1953 – 1964

1969 – 1977

Pu and Np and some Tc-99 

were concentrated in ash 

and residue.

Potential for worker 

exposure to 

concentrated RU 

constituents.

Pulverization of tower ash 

and other recovery 

processes. 

C-410,

C-420

1953 – 1964

1968 – 1976 

Pu and Np and some Tc-99 

were concentrated in ash 

and residue.

Potential for worker 

exposure to 

concentrated RU 

constituents.

Recycling of U in ash through 

fluorination. 

C-410,

C-420

1953 – 1964

1969 – 1977

Pu and Np and some Tc-99 

were concentrated in ash 

and residue.

Potential for worker 

exposure to 

concentrated RU 

constituents.

Packing of waste ash for 

storage prior to shipping. 

C-410,

C-420

1953 – 1964

1969 – 1977

Pu and Np and some Tc-99 

were concentrated in ash 

and residue.

Potential for worker

exposure to 

concentrated RU 

constituents.

Transfer of UF6 to cylinder 

from fluorination process. 

C-410,

C-420

1953 – 1964

1969 – 1977

Pu and Np and some Tc-99 

became further 

concentrated in cylinder 

heels.

Withdrawing UF6 from 

cylinder and feeding into 

cascade. 

Cascade feed 

points. 

1953 – 1964

1969 – 1970

1972 - 1976

Pu and Np and some Tc-99 

were concentrated in 

cylinder heels. Np and 

smaller amounts of Pu 

exiting cylinder with UF6

collected at feeder head 

leading to cascade and in 

cascade near feed point.

Venting to atmosphere from 

cascade. 

Cascade 

purge 

locations; 

C-331, 

C-333, 

C-335, 

C-337, 

C-310

1952 - 1985 As RU moved higher in 

cascade, Tc-99 became 

concentrated, but Pu and 

Np were much diminished. 

Venting at higher points in 

cascade involved 

concentrated Tc-99.

Traps were used to 

remove Tc-99 before 

venting.

CIP/CUP equipment removal 

and other equipment removal, 

maintenance, and 

decontamination. 

Cascade 

buildings 

C-270 Scrap 

Yard Burial 

Grounds. 

1954 - 1962

1973 - 1981

Np and lesser amounts of 

Pu collected near cascade 

feed points.

Removal, handling, 

and decontamination 

of equipment near feed 

points could have 

exposed workers to 

concentrations of small 

percentages of Np and 

Pu that actually 

entered cascade.
Document No. ORAUT-TKBS-0019-2 Revision No. 01 Effective Date: 08/21/2006 Page 15 of 29

NOTE: A complete list of radionuclides, quantities, chemical compounds, and inhalation class is not 

provided for every PGDP process and building listed above. ORAUT (2005a) provides a more 

complete description of this information.

2.4.2

In Building C-310, enriched UF6 product was removed from the cascade and put in cylinders for 

transport. Building C-310 was approximately 53 ft by 30 ft in area and had two roll-up doors, one 

employee access door, and double doors to the storage room. The building was equipped to handle 

two 10- to 14-ton cylinders at a time.

C-310 – Purge and Product Withdrawal Building

Building C-310 received enriched UF6 gas from the cascades product via pumps that discharged 

through a condenser, piping, and cylinder pigtails to the intended receiving UF6 cylinder. Product 

cylinders were filled to no more than 95% (liquid) of capacity.

Manual valves on the UF6 cylinder would occasionally be identified as defective and require 

replacement. This work was performed in Building C-310. According to procedures in effect in the

1970s, a UF6 cylinder was required to cool at least 5 d before its valve was replaced. Those cylinders 

known to be above atmospheric pressure after the minimum cooling period would be vented and 

further cooled, if necessary, with cold water. Until the mid-1970s, defective UF6 cylinder valves were 

routinely replaced by the mechanic standing upwind, with any escaping gases or fumes going the 

other way.

In 1980, the building was normally staffed by one to three workers with a crane operator on call if 

cylinder transfer involving crane movements was required. The workers were responsible for 

completing equipment checks, logging equipment data, preparing cylinders for filling, disconnecting 

and weighing full cylinders, transferring cylinders, and maintaining cylinder records. In 1997, the 

workforce consisted of operators (three to seven per shift), maintenance mechanics (four per shift), 

instrument mechanics (two per shift), electricians (two per shift), and foremen (four per shift).

Operations Performed in Building C-310 and Maximum Radionuclide Concentrations Expected

(Bechtel Jacobs 2001):

Connecting and disconnecting UF6 product cylinders and handling product UF6 – 1953–1999

27 ppb 237Np

0.06 ppb 239Pu

20,000 ppb 99Tc (UF6) 

Minimal external radiation exposure potential

Changing/cleaning MgF2 traps – 1964–1966

1,900 ppb 237Np

0.01 ppb 239Pu

2,266,000 ppb 99Tc (MgF2) 

Moderate external radiation exposure potential

2.4.3

Building C-315 began operation in early 1953. The building was used for the removal of depleted UF6

byproduct from a cascade and storage in cylinders. It was approximately 53 ft by 30 ft in area and 

contained four cart tracks and product equipment to accommodate four 10- to 14-ton cylinders. Four 

roll-up doors in the east wall permitted the entry and exit of cylinders, while the west wall contained 

C-315 – Surge and Tails Withdrawal Building
Document No. ORAUT-TKBS-0019-2 Revision No. 01 Effective Date: 08/21/2006 Page 16 of 29

doors to the pump room and control room. Thus, six penetrations (doors) affected air flow in the 

building.

Liquefaction was accomplished by compression of UF6 piped to the building from the enrichment 

operation (Buildings C-331, C-333, C-335, and C-337) at a pressure at which UF6 gas can be 

conveniently liquefied. After condensing, the liquid was allowed to flow into cylinders. The product 

was drained as a liquid into the multiton cylinders through a copper tube referred to as a pigtail. 

When the cylinder was filled to capacity, the cylinder and drain valves were closed and the pigtail was 

evacuated and purged. The pigtail was disconnected at the cylinder valve (PACE/University of Utah 

2000).

In 1980, the building was normally staffed by one to three workers with a crane operator on call if 

cylinder transfer involving crane movements was required. The workers were responsible for 

completing equipment checks, logging equipment data, preparing cylinders for filling, disconnecting 

and weighing full cylinders, transferring cylinders, and maintaining cylinder records. By 1997, the 

workforce consisted of operators (two or three per shift), maintenance mechanics (four per shift), 

electricians (two per shift), janitor (one per shift), and foremen (four per shift) (PACE/University of 

Utah 2000).

Operations Performed in Building C-315 and Maximum Radionuclide Concentrations Expected

(Bechtel Jacobs 2001):

Connecting and disconnecting UF6 tails cylinders and handling tails UF6 – 1953–1999

5 ppb 237Np

0.1 ppb 239Pu

20,000 ppb 99Tc (UF6) 

Minimal external radiation exposure potential

2.4.4

The main process buildings at PGDP contain the cascades, which are series of compressor and 

converter stages and supporting equipment arranged in units and cells that progressively enrich UF6

in its gaseous form. The process buildings encompass 74 acres under roof. Each process building 

has 1,812 enrichment stages.

Gaseous Diffusion Process Buildings (C-331, C-333, C-335, and C-337)

PGDP construction, from 1951 through 1956, occurred in two phases. Construction of the first phase 

began January 2, 1951, and included Buildings C-331 and C-333. Authorization to proceed with the 

second phase of PGDP construction was received on July 15, 1952. Two additional enrichment 

facilities, Buildings C-335 and C-337, were added, and construction was completed in 1956. In 

September 1952, Buildings C-331 and C-333 began operation. In November 1952, the first product 

was withdrawn. In July 1953, the first reactor tails of UF6 were fed to the enrichment cascade after 

conversion in Oak Ridge. In April and July 1954, Buildings C-335 and C-337, respectively, began 

operations (Bechtel Jacobs 2001). 

Operations Performed in the Cascade Buildings and Maximum Radionuclide Concentrations 

Expected (Bechtel Jacobs 2001):

Cascade operations – 1953–1964, 1969–1970, and 1972–1976

450 ppb 237Np

0.09 ppb 239Pu

23,000 ppb 99Tc

Moderate external radiation exposure potential
Document No. ORAUT-TKBS-0019-2 Revision No. 01 Effective Date: 08/21/2006 Page 17 of 29

Cascade maintenance (cascade dust) – 1954–1961 and 1973–1981 (CIP/CUP)

2,740 ppb 239Pu

3,220,000 ppb 237Np

Specific concentrations of 99Tc not available for this operation

High external radiation exposure potential

Isotopic concentrations in the reactor tails (approximate):

0.2 ppm 237Np

4 ppb 239Pu

7 ppm 99Tc

2.4.5

Several operations were performed in this building, two of which had a high potential for increased 

worker radiation exposure. These were the conversion of depleted UF6 to UF4 using a hydrogenation 

process and the conversion of some UF4 to uranium metal via a reaction with magnesium. The 

rationale for doing hydrogenation was to recover hydrofluoric acid for use in the oxide conversion 

process in Building C-420. Another reason might have been to convert UF6 to a form that was easier 

to store. Both processes generated considerable amounts of dust.

Building C-340 – Metals Plant

Metals production involved several steps, each with its own hazards:

• The uranium metal production process involved reducing UF6 (normally from the tails cylinder) 

to UF4 by combining it with hydrogen in a heated tower. The UF4 was mixed with magnesium 

and fed into lined firing reduction vessels, placed in furnaces, and heated until it fired into a 

metal ingot, called a derby. The next phase of the operation involved blending measured 

quantities of UF4 (depleted uranium) with measured quantities of magnesium metal and 

pouring this mixture into the reactor liner. A refractory cap was then poured, and a lid was 

bolted to the top of the charged reactor. The charged reactor was transferred to an induction 

furnace where it was heated to the point where the uranium reduction started.

The primary hazard associated with this part of the process was exposure to airborne uranium 

dust during weighing, blending, and pouring. Respirators were required soon after initial 

production operations began. In addition, the reactors presented a significant hazard from 

burning magnesium and molten uranium metal. Phenomena described as "burnout" and "lid 

fires" occurred infrequently when the refractory liner was not correctly prepared. For example, 

burnouts occurred when burning magnesium came in contact with the steel shell, melting 

through the shell and releasing reactor contents to the furnace. Lid fires were similar, but 

occurred at the lid rather than the side of the shell. Such an occurrence led to a fatality (death 

was not due to radiological exposure) in March 1962. Burnouts resulted in significant 

contamination of the furnace refractory and would normally require relining of the entire 

furnace.

After the reactor was cooled, it was sent to the breakout area where the lid was removed, the 

shell was inverted, and the contents were dumped on a grating, referred to as a grizzly. The 

slag material, at this point a hard ceramic, was broken into smaller pieces by beating it with a 

hammer. The pieces dropped through the grating into a jaw crusher and went to the slag 

plant. This operation was among the dirtiest jobs in Building C-340. Operators were 

sometimes covered with black dust. Although respirators were required and generally worn, 

the extent of dust and contamination might have exceeded the protection factor of the 

respirator.
Document No. ORAUT-TKBS-0019-2 Revision No. 01 Effective Date: 08/21/2006 Page 18 of 29

• The metal ingot, or derby, was freed from the slag and roasted to oxidize the surface and 

loosen remaining slag. Loose oxides that fell from the derbies during roasting were collected, 

put in drums, and sent to a burial yard.

• After roasting, the derbies were cleaned by hand in a cleaning booth using power brushes and 

grinders to remove remaining slag. The potential hazards for airborne contamination for this 

operation were similar to those for the breakout operation.

• After cleaning, the derbies could be shipped directly or sawed into smaller shapes, depending 

on customer requirements. Derby sawing generated large amounts of uranium metal 

"sawdust," which burns readily in air. This sawdust was collected in drums of oil and kept 

covered. Despite these measures, uranium metal fires were common (daily or weekly), 

resulting in elevated levels of airborne uranium oxides.

• The Building C-340 operation was capable of remelting uranium derbies and casting specific 

shapes. Operations were conducted in a furnace with a controlled atmosphere. Graphite 

crucibles were used to receive molten uranium. The primary hazard associated with these 

operations was cleaning the crucibles between pours. Over time, oxides of uranium and 

beta-emitting uranium decay products would impregnate a crucible. Because crucibles were 

cleaned by hand, operators received radiation doses to their hands, arms, and fingers.

• The MgF2 reaction product remaining in the reactor was captured, crushed, ball milled, and 

sized to be recycled as refractory. Although this was primarily a hands-off operation, it 

generated significant quantities of dust. Over time, the slag became contaminated with 

significant quantities of uranium oxides (several percent) that could have contributed to worker 

intakes. Reject slag (too small or too large) was collected in a hopper, and periodically 

drummed. Operators did not wear dosimetry that would have measured these extremity 

exposures (see ORAUT 2005b).

From 1978 to 1982, Building C-340 served as a shipping point for UF4. In addition, the building 

contained a remelt furnace for recasting uranium. The workforce consisted of operators (10 to 20 per 

shift), maintenance mechanics (three to five per shift), instrument mechanics (three to five per shift), 

and electricians (three to five per shift) (PACE/University of Utah 2000). After enrichment operations 

were shut down in the late 1970s, the building was used for other activities (e.g., offices and training 

programs).

Operations Performed in Building C-340 and Maximum Radionuclide Concentrations Expected

(Bechtel Jacobs 2001):

Production and handling of UF4 produced from tails – 1957–1962 and 1967–1977

62.6 ppb 237Np

0.063 ppb 239Pu

49 ppb 99Tc (UF4) 

Minimal external exposure potential

Connecting and disconnecting UF6 tails cylinders and handling UF6 tails – 1953–1999

5 ppb 237

Np

0.1 ppb 239Pu

20,000 ppb 99Tc (UF6) 

Minimal external exposure potential
Document No. ORAUT-TKBS-0019-2 Revision No. 01 Effective Date: 08/21/2006 Page 19 of 29

Uranium metal production from metal production, sawing, and metal handling – 1957–1962 and 

1967–1977

2.7 ppb 237Np

0.0085 ppb 239Pu

10.6 ppb 99Tc (uranium metal)

Minimal external exposure potential

Handling MgF2 in uranium metal manufacturing – cleaning, roasting, liner preparation, knock out, 

and drumming – 1957–1962 and 1968–1977

2,400 ppb 237Np

1.2 ppb 239Pu

1,880 ppb 99Tc (MgF2) 

High external exposure potential

2.4.6

Building C-400 was brought into service in April 1953. Its uranium recovery facilities were used to 

chemically separate and recover uranium from a variety of waste materials. Sources of feed material 

for this process included fluorination tower ash, sintered metal filters, decontamination solutions, UF6

scrubber solutions, particulates from ventilation filters and vacuum cleaners, laboratory wastes, and 

materials from spills. Building C-400 was a site for potentially increased radiation exposures, primarily 

because the following operations occurred there: converter disassembly, pulverization of waste UF4

and recycled UO3 containing TRU elements, cylinder heel cleaning, spray booth operation, and 237Np 

and 99Tc recovery. In addition, radiological hazards were associated with cleaning the building air 

filtering system (baghouses) (Bechtel Jacobs 2001). 

Building C-400 – Decontamination and Cleaning Building/Uranium Recovery

Before the mid-1970s, a complex uranium recovery process was performed in Building C-400. 

Uranium was separated from waste and scrap materials, concentrated, and converted to an oxide. 

The process included the following steps: dissolution of feed materials, filtration, solvent extraction in 

pulse columns, concentration by evaporation, and denitration to an oxide. The aqueous raffinate 

waste from solvent extraction columns, which contained 237Np, 239Pu, and 99Tc, was discharged to the 

environment (Bechtel Jacobs 2001). 

In the mid-1970s, the solvent extraction process for uranium recovery was replaced with a simpler 

precipitation and filtration process. Steps in this new process included dissolution of feed materials in 

nitric acid, addition of lime to precipitate uranium, and recovery of precipitated uranium as a filter 

cake.

Maintenance on major components in the cascade (compressors, converters, and process block 

valves) presented some of the most significant opportunities for exposure of maintenance personnel. 

Work on these components required that they be removed from the system, cleaned, rebuilt or

repaired, and reinstalled. To remove these components, process operators isolated and bypassed 

the cascade cell containing the component, reduced the UF6 in the cell to less than 10 ppm equivalent 

at atmospheric pressure, and purged the cell to minimize HF and UF6 exposure to workers involved in 

opening, maintaining, or modifying cell components. Once a satisfactory UF6 negative and HF purge 

were accomplished and the pressure of the isolated cell was raised to atmospheric pressure with dry 

air, that cell was turned over to process maintenance for opening and disassembly.

Compressors were transported from the process buildings to Building C-400 (and Building C-720). 

The compressors were disassembled into major components in pits, the parts were spray-washed to 
Document No. ORAUT-TKBS-0019-2 Revision No. 01 Effective Date: 08/21/2006 Page 20 of 29

remove uranium deposits, and the rotor was relocated as required for deblading in Building C-400. 

The barriers were washed and disassembled, and scrap recovery was performed.

Neptunium Recovery

Soon after 237Np was identified at Paducah in 1957, the AEC placed a high emphasis on its recovery. 

A 237Np recovery process developed at Oak Ridge National Laboratory (ORNL) began operation at 

PGDP in November 1958 in Building C-400. The process used a solvent extraction and evaporation 

method to recover and concentrate 237Np from receiver ash and cylinder heels wash solution. 

Receiver ash and solids that settled from cylinder washwater were dissolved in a nitric acid solution. 

Solids suspended in this solution were removed by filtration and discarded as solid waste. The filtrate 

was processed through solvent exchange pulse columns to separate uranium, thorium, and 237Np. 

These columns, originally in Building C-710, were moved to Building C-400 sometime after July 1959. 

Raffinate from these columns was dumped to the building drain if it contained uranium and 237Np 

concentrations less than 500 ppm and 2.0 mg/L, respectively. Uranium and thorium were recovered 

for future use. The 237Np solution was concentrated to about 20 to 25 g/L by evaporation (Bechtel 

Jacobs 2001). 

The relatively greater radiological hazards associated with 237Np were understood at PGDP as early 

as 1959, and special practices for handling 237Np solutions and 237Np-contaminated equipment were 

recommended. Recommendations included using nonbreakable containers; maintaining tight 

systems; keeping lids on containers; preventing bubbling, frothing, or spraying of solutions; using 

rubber gloves; washing the gloves before using them in other areas; using respirators (or assault 

masks) for welding or burning; and performing alpha surveys of all equipment removed from 237Np 

processing areas (Bechtel Jacobs 2001). 

In 1958, ash and cylinder washings were processed through the Building C-400 237Np recovery 

process. The process used aqueous chemistry and ion exchange methods to recover 3.215 kg of 

237Np from the cylinder wash stream and 1.074 kg of 237Np from the ash stream. The recovered 

materials were shipped to the Hanford Site. A program to recover 99Tc from cylinder washwater and 

raffinate (e.g., solvents) from 237Np recovery operations began in April 1960 (Bechtel Jacobs 2001).

Technetium Recovery

Technetium-99 is a fission product that PGDP received in recycled feed from the Hanford and 

Savannah River Sites. Technetium-99 passed through the cascade as a volatile compound of 

fluorine, depositing on internal surfaces of the cascade and contaminating the enriched uranium 

product. The AEC did not specify a limit for 99Tc in UF6 feed but indirectly controlled its concentration 

to about 10 ppm by limiting gross beta from fission products. A demand for 99Tc in the early 1960s 

prompted PGDP to begin a process to recover 25 kg of this material from various effluent streams. In 

1960, a process was begun to recover 99Tc from UF6 cylinder washwater and from raffinate generated 

during 237Np recovery. Process steps included precipitation and removal of uranium from these 

solutions by adding sodium hydroxide. This solution was processed through an ion exchange column 

and elutriated with nitric acid to produce a concentrated solution of 99Tc that was shipped to ORNL. 

Although the contribution to radiation dose from 99Tc was not of concern during most PGDP operation 

and maintenance activities, the isolated activity required specific monitoring considerations for both 

internal exposure and external skin exposure (Bechtel Jacobs 2001). 

Cylinder Cleaning

With repeated reuse, UF6 cylinders collected deposits that did not completely volatilize in the 

autoclave. These deposits, called cylinder heels, had to be dissolved and removed periodically, and 

the cylinder was then cleaned, refurbished as necessary, reinspected, hydrostatically tested, and 

weighed for subsequent use. Cylinder heels were composed of corrosion products, uranium salts and 
Document No. ORAUT-TKBS-0019-2 Revision No. 01 Effective Date: 08/21/2006 Page 21 of 29

oxides, and TRU and uranium daughter product compounds. With regard to contaminants of the 

process gas, some of the 237Np and much of the 99Tc was volatilized to the cascade, while most 239Pu 

remained in the cylinder heels, creating a significant radiological hazard. Cylinder cleaning was 

performed at Building C-400, where the heels were dissolved and the rinse water was collected in a 

large pan (Bechtel Jacobs 2001). 

Operations Performed in Building C-400 and Maximum Radionuclide Concentrations Expected

(Bechtel Jacobs 2001):

Cascade maintenance - 1954–1961 and 1973–1981 (CIP/CUP)

450 ppb 237Np

0.09 ppb 239Pu

Specific concentrations of 99Tc not available for this operation

High external radiation exposure potential

Uranium/neptunium recovery – 1958 to the late 1970s 

25,000,000 ppb 237Np

100,000 ppb 239Pu

Specific concentrations of 99Tc not available for this operation (solutions)

Moderate external radiation exposure potential

2.4.7

This area was the primary disposal site for radioactive waste at PGDP. It was constructed as a 

holding pond for Building C-400 liquid waste. The pond was 380 ft by 140 ft, with 6-ft-high dikes. In 

1957, the C-404 Holding Pond was converted to a solid radioactive waste burial area. By 1977, 

approximately 6,400,000 lb of materials contaminated with uranium had been drummed and placed in 

the holding area. Waste streams included incinerator ash, contaminated alumina, highly 

contaminated roofing waste, and gold recovery sludge. This area continued in use into the mid1980s. It was later determined to contain sludge that was chemically hazardous, requiring closure 

under the Resource Conservation and Recovery Act in 1987 (Bechtel Jacobs 2001). 

Building C-404 – Solid Radioactive Waste Disposal Area

Reported activities or concentrations for other radionuclides were not found for this area.

2.4.8

Cascade Maintenance

Building C-409 – Stabilization Building

Once compressors had been overhauled and reassembled in Building C-400 or C-720, compressor 

openings were covered for transportation to storage or reinstallation. Converters were transported 

from the process buildings to Building C-409 for decontamination.

During the 1973–1981 cascade improvement, a shop for reassembling and testing converters was 

located in Building C-409. In addition, the building had a small spray booth for minor cleaning jobs. 

The number of personnel involved is not known. The spray booth was used in the initial stages of 

tearing down converters. The waste from this booth was piped to Building C-400 for removal of 

uranium oxides. It is not clear, but apparently the Building C-409 spray booth was operated by 

Building C-400 personnel.

Operations Performed in Building C-409 and Maximum Radionuclide Concentrations Expected

(Bechtel Jacobs 2001):

Cascade maintenance – 1954–1961 and 1973–1981 (CIP/CUP)

450 ppb 237Np
Document No. ORAUT-TKBS-0019-2 Revision No. 01 Effective Date: 08/21/2006 Page 22 of 29

0.09 ppb 239Pu

Specific concentrations of 99Tc not available for this operation

High external radiation exposure potential

2.4.9

From July 1953 through 1977, UF6 feed material was produced from UO3 (yellowcake) at PGDP in 

Buildings C-410 and C-420. This feed material was supplied by sources such as El Dorado Mining 

and Refining, Mallinckrodt Chemical Works, and Allied Chemical and comprised less than 10% of the 

UF6 fed to the cascade. From 1953 through 1964 and intermittently from 1968 through 1977, the feed 

plant produced UF6 from UO3 obtained from spent reactor fuel processed at the Hanford and 

Savannah River Sites. After 1977, all feed came in the form of UF6 from outside sources such as Oak 

Ridge, Portsmouth, and Allied Chemical.

Building C-410 – Feed Plant

The next stage of the enrichment operation was to convert solid UF4 to gaseous UF6. This was done 

using heated elemental fluorine gas in the Building C-410 fluoridation towers. The operation 

consisted of introducing UF4 at the top of the tower while fluorine gas was introduced from below. The 

resulting UF6 gas/liquid was removed in large cylinders and the solid waste products were collected in 

ash receivers at the bottom of the towers. The external radiological concerns were from beta and 

gamma emissions from TRU elements, fission products, and accumulated uranium daughter 

products, which were concentrated by the process in the ash. In addition to external radiation 

sources, inhalation of radioactive dust while cleaning plugged equipment, changing the ash receivers, 

and cleaning the building air filters was possible. The potential for radiation exposure was particularly 

increased for work around and with ash receivers. Workers in the fluorination tower area were 

exposed to UF4 and uranyl fluoride UO2F2 (PACE/University of Utah 2000).

Exposure to uranium powder dusts was prevalent in operations and maintenance activities. For 

example, the plugging of conveyers, hoppers, and screws with UO3 or UF4 routinely required physical

agitation with sledgehammers or metal rods. In many cases, shear pins or chains on the associated 

drive mechanisms broke, requiring operations personnel to clean the product out of the jammed 

equipment and maintenance personnel to disassemble and repair the equipment. 

Operations Performed in Building C-410 and Maximum Radionuclide Concentrations Expected

(Bechtel Jacobs 2001):

Cascade operations – 1953–1964, 1969–1970, and 1972–1976

450 ppb 237Np

0.09 ppb 239Pu

23,000 ppb 99Tc

Moderate external radiation exposure potential

Feed plant operations, RU to UF6 - UO3 → UO2 UF4 UF [handling, drumming, bag changing, 

etc. (tower ash)] – 1953–1964 and 1969–1977

25,602 ppb 237Np

8,000 ppb 239Pu

4,600 ppb 99Tc

High external radiation exposure potential

Changing/cleaning MgF2 traps – 1964–1966

1,900 ppb Np

0.01 ppb Pu

2,266,000 ppb Tc (MgF2) 
Document No. ORAUT-TKBS-0019-2 Revision No. 01 Effective Date: 08/21/2006 Page 23 of 29

Moderate external radiation exposure potential

2.4.10

In August 1956, the Building C-420 expansion to the feed plant was complete. Building C-420 was a 

comparatively small building attached to the west side of Building C-410. It contained fluidizing beds 

used in the conversion processes. This was where triuranium octaoxide (U3O8) was converted to 

uranium oxide and then to UF4 for use as feed stock for the fluorination towers in Building C-410. The 

ore ran through two sets of fluidized beds, the first of which converted UO3 (yellowcake) to UO2 (black 

powder), while the second converted UO2 to UF4 by reaction with hydrofluoric acid. The primary 

potential for radiation exposure to operators was the inhalation of dust generated while unplugging 

converters or while cleaning the building air filtering system (baghouse). Maintenance mechanics had 

a potential for inhalation exposure while working on the equipment.

Building C-420 – Oxide Conversion Plant (UF4 – Green Salt Plant)

The equipment used for the conversion processes consisted of a series of hoppers, conveyer belts, 

screws, chutes, etc., which were susceptible to mechanical failure. When a failure occurred, the 

operators and maintenance mechanics would open the system and the operators and maintenance 

mechanics did what was necessary to resume operations. Routine operations in Building C-420 do 

not appear to have had the potential for increased radiation exposure; however, workers in this 

building were rotated with those in Building C-410 to reduce individual radiation exposure.

The workforce consisted of operators (four per shift), maintenance mechanics (two per shift), 

electricians (two per shift), instrument mechanics (two per shift), and janitors (one per shift) (PACE/ 

University of Utah 2000).

Operations Performed in Building C-420 and Maximum Radionuclide Concentrations Expected

(Bechtel Jacobs 2001):

Feed plant operations, RU to UF6 - UO3 → UO2 UF4 UF (handling, drumming, bag changing, etc. 

(tower ash) - 1953–1964, 1969–1977, and 1982–1983

25,602 ppb 237Np

8,000 ppb 239Pu

4,600 ppb 99Tc

High external radiation exposure potential

2.4.11

The PGDP analytical laboratories are in Building C-710. These facilities, which have been in 

operation since plant startup, consist of the American Society of Testing and Materials, Industrial 

Hygiene, Infra-Red Spectrometry, Emission Spectrometry, Sampling, Uranium Analysis, Trouble 

Shooting, Quality Control, Radiochemistry, Metallurgy, Mass Spectrometry, Counting Preparation, 

Alpha and Beta Counting, and Fission Training Laboratories.

Building C-710 – Analytical Laboratories

The 99Tc recovery operation, initially in Building C-710, moved to Building C-400 after 1959.

A 237Np recovery process originally in Building C-710, Room 32, transferred to Building C-400 

sometime after July 1959 (see Section 2.4.6). After processing in Building C-400, the concentrate 

went to a laboratory in Building 710 for additional separation and concentration in ion exchange 

columns. The final product was siphoned into glass carboys on the Building C-710 loading dock and 

shipped to ORNL.
Document No. ORAUT-TKBS-0019-2 Revision No. 01 Effective Date: 08/21/2006 Page 24 of 29

A Health Physics and Hygiene report (Paducah 1959) for the first quarter of 1959 stated that 

continuous air samples collected near the 237Np recovery operation in Building C-710 averaged 

slightly above the maximum air concentration, a term used for airborne radioactivity concentrations 

assumed for 237Np. Later analysis indicated that 29% of the alpha activity was attributable to 237Np. 

There is no indication that respiratory protection was used during these activities. Urine samples that 

were collected and sent to ORNL for analysis tested positive for 237Np (PACE/University of Utah 

2000).

Operations Performed in Building C-710 and Maximum Radionuclide Concentrations Expected

(Bechtel Jacobs 2001):

Uranium/neptunium recovery - 1958 to late 1970’s

25,000,000 ppb 237Np

100,000 ppb 239Pu

Specific concentrations of 99Tc not available for this operation (solutions)

Moderate external radiation exposure potential

2.4.12

Building C-720 contained the compressor, converter, and machine shops. Machinists, maintenance 

mechanics, instrument mechanics, sheet metal workers, electricians, and inspection workers 

performed the following functions: disassembly of compressors, converter maintenance, disassembly 

of block valves, machining, fabrication, welding, and grinding. In addition, stores workers and janitors 

were assigned to the building. Supervisory offices were in the middle of the building.

Building C-720 – Maintenance Building

The primary structure of interest from the standpoint of radiation safety was the compressor 

disassembly area. This was in a pit at one end of the building and was several stories high. 

Occasional releases of UF6 occurred during compressor disassembly.

Compressors were transported from the process buildings to Building C-720. The compressors were 

disassembled into major components in pits, the parts were transported to Building C-400 for spray 

washing to remove uranium deposits, and the rotor and stator were relocated as required for 

deblading in Buildings C-400 and C-410, respectively. Reusable washed parts were returned to their 

respective maintenance buildings for modification, refurbishment, degreasing, and reassembly. 

Following washing in Building C-400, the converters were modified, refurbished, and reassembled in 

Building C-720. Prior to removal from the system, block valves were slightly opened (where possible), 

inspected, cut out of the system, lifted free of process piping, decontaminated, and, after covers were 

installed, shipped to Building C-400 for preliminary disassembly and decontamination to the limits 

allowed in Building C-720. Once decontaminated, the valve was covered and transported to 

Building C-720 for final repair and reassembly, and staged in the process building for reinstallation.

Support operations involved the following exposure pathways:

• Cleaning of cylinder heels (potentially involving feed, product, or tails cylinders)

• Decontamination of equipment associated with feed, cascade, and other operations

• Routine and emergency maintenance operations at ancillary support facilities

• Uranium recovery from oils, cleaning solutions, and other wastes

• Effluent, sludge, and other wastes from decontamination processes

• Incineration of certain wastes

• Scrap from equipment

• Removal or drainage of sludge from waste ponds

• Analytical laboratory sampling
Document No. ORAUT-TKBS-0019-2 Revision No. 01 Effective Date: 08/21/2006 Page 25 of 29

Operations Performed in Building C-720 and Maximum Radionuclide Concentrations Expected

(Bechtel Jacobs 2001):

Cascade maintenance -- 1954–1961 and 1973–1981 (CIP/CUP)

450 ppb 237Np

0.09 ppb 239Pu

Specific concentrations of 99Tc not available for this operation

High external radiation exposure potential
Document No. ORAUT-TKBS-0019-2 Revision No. 01 Effective Date: 08/21/2006 Page 26 of 29

REFERENCES

Bechtel Jacobs (Bechtel Jacobs Company, LLC), 2001, Recycled Uranium Mass Balance Project 

Paducah Gaseous Diffusion Plant Site Report, BJC/PGDP-167 DRAFT, prepared for the DOE 

Office of Environmental Management.

Lockheed Martin (Lockheed Martin Energy Systems, Inc.), 1997a, Paducah Site Annual 

Environmental Report for 1996, Document No. KY/EM-206, prepared for the U.S. Department 

of Energy, Kevil, Kentucky, December.

Lockheed Martin (Lockheed Martin Energy Systems, Inc.), 1997b, Paducah Site Annual 

Environmental Report for 1995, Document No. KY/EM-176, prepared for the U.S. Department 

of Energy, Kevil, Kentucky, January.

ORAUT (Oak Ridge Associated Universities Team), 2005a, Technical Basis Document for Paducah 

Gaseous Diffusion Plant – Occupational Internal Dose, ORAUT-TKBS-0019-5, Rev. 00-PC1-B, Oak Ridge, Tennessee.

ORAUT (Oak Ridge Associated Universities Team), 2005b, Technical Basis Document for Paducah 

Gaseous Diffusion Plant – Occupational External Dose, ORAUT-TKBS-0019-6, Rev. 00, Oak 

Ridge, Tennessee.

PACE/University of Utah, 2000, Exposure Assessment Project at the Paducah Gaseous Diffusion 

Plant, submitted by Paper, Allied Industrial, Chemical and Energy Workers (PACE) 

International Union and University of Utah, Division of Radiobiology, Center for Advanced 

Medical Technologies, Center of Excellence in Nuclear Technology, Engineering and 

Research.

Paducah, 1959. Quarterly Report – January, February, March, 1959, Health Physics and Hygiene 

Report, Author not given, SRDB No. 18755
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GLOSSARY

absorption type 

Categories for materials according to their speed of absorption in the body, which replaced the 

earlier inhalation clearance classes. Defined by the International Commission on Radiological 

Protection, the absorption types are F for fast absorption (formerly inhalation class D), M for 

moderate absorption (formerly inhalation class W), and S for slow absorption (formerly 

inhalation class Y). Also called solubility type. See inhalation class. 

cascade

At PGDP, series of compressor, heat exchanger, control valve and motor, converter stages, 

and supporting piping arranged in stages, cells, and units that progressively increase the 

concentration of 235U in a uranium hexafluoride (UF6) feed. Enrichment occurs as UF6 passes 

through semiporous barriers in the converter stage. These barriers allow the lighter 235U 

molecules to pass through more easily, which results in a gas with a slightly higher percentage 

of 235U (enriched) on one side of the barrier and a slightly lower percentage (depleted) on the 

other side. The enriched UF6 gas flows toward the top of the cascade while the depleted UF6

gas travels toward the bottom of the cascade.

contamination control program

System of controls to reduce and confine radioactive contamination. These systems include 

contamination surveys, posting, containment devices, and the prevention, detection, and 

timely repair of leaks. 

depleted tails

Material that has passed through one of the cascade stages that has been depleted of some 

of its 235U.

derby

At PGDP, molded uranium metal ingot of UF4 and magnesium.

dosimetry

Measurement and calculation of internal and external radiation doses.

enriched uranium

Uranium in which processing has increased the proportion of 235U to 238U to above the natural 

level of 0.7%. Reactor-grade uranium is usually about 3.5% 235U; weapons-grade uranium 

contains greater than 90% 235U.

external radiation 

Radiation from sources outside the body.

feed material

At PGDP, UF6 used in the cascade systems for the enrichment process.

fission product

Radionuclide produced by fission or by the subsequent radioactive decay of radionuclides.

gaseous diffusion enrichment 

Process by which uranium hexafluoride (UF6) is filtered through a series of semipermeable 

molecular barriers to separate the lighter 235U from the heavier, more naturally abundant 238U. 
Document No. ORAUT-TKBS-0019-2 Revision No. 01 Effective Date: 08/21/2006 Page 28 of 29

High Radiation Area

Area with dose rates greater than 100 millirem (1 millisievert) in 1 hour at 30 centimeters from 

the source or its container. HRAs require posted signs and strict access control.

inhalation class

The respiratory tract inhalation classification scheme developed in ICRP Report 30 for inhaled 

material according to its rate of clearance from the pulmonary region of the respiratory tract

(ICRP 1979). Materials are classified as D (days), W (weeks), or Y (years), according to how 

fast they clear the lungs; Class D in less than 10 days; Class W in 10 to 100 days; Class Y in 

more than 100 days. More recent recommendations in ICRP Report 66 replaced classes D, 

W, and Y with lung absorption Types F (fast), M (moderate), and S (slow), respectively (ICRP 

1994).

internal radiation exposures

Radiation exposure absorbed by the body resulting from radioactive material taken into the 

body.

liquefaction

Process of causing a gas or solid to become a liquid, usually through condensing a gas or 

melting a solid.

natural uranium

Uranium as found in nature, approximately 99.27% 238U, 0.72% 235U, and 0.0054% 234U by 

weight. The specific activity of this mixture is 2.6 × 107 becquerel per kilogram (0.7 picocuries 

per gram).

pigtail

At PGDP, a flexible connecting device for draining depleted liquefied UF6 gas into storage 

cylinders.

production reactor tails

At PGDP, uranium reprocessed from spent reactor fuel used in a plutonium production reactor 

(e.g., at the Hanford and Savannah River Sites).

reactor tails

At PGDP, uranium reprocessed from spent reactor fuel, normally from production reactors, but 

some from reactors in foreign countries.

reactor vessel

At PGDP, hollow receptacle in which a chemical reaction takes place.

refractory cap

At PGDP, cover of highly heat-resistant material underneath the reactor vessel lid.

roasting

At PGDP, process of heating derbies to remove all the slag produced during the reduction 

process (heating UF4 and magnesium to form uranium derbies).
Document No. ORAUT-TKBS-0019-2 Revision No. 01 Effective Date: 08/21/2006 Page 29 of 29

spent reactor fuel

Reactor fuel that has been used in a reactor to the point that the amount of fissionable 

uranium (235U) has been depleted.

transuranic (TRU) elements

Elements with atomic numbers above 92 (uranium). Examples include plutonium and 

americium. All isotopes of the transuranic elements are radioactive, they are naturally either 

rare or nonexistent on Earth, and most are known only as a result of research using nuclear 

reactors and particle accelerators because of extremely short half-lives. 

weekly plant action level

At PGDP, radiation exposure or airborne radioactivity value that, when reached by an 

individual, would result in some compensatory action being taken for that individual, such as 

reassigning the individual to a work detail involving no or reduced radiation exposure or 

airborne radioactivity.